Text

Technical Specification Proposed Change No. 276 Revision to TS Sections 3.3/4.3 for Control Rod System
	ML080430607
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	02/06/2008
From:	Ted Sullivan; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 08-007
	Download: ML080430607 (63)
	v • d • e

Entergy Entergy Nuclear Operations, Inc.

Vermont Yankee P.O. Box 0250 320 Governor Hunt Road Vernon, VT 05354 Tel 802 257 7711 February 6, 2008 BVY 08-007 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

REFERENCE:

SUBJECT:

1)

NUREG 1433, Revision 3, "Standard Technical Specifications General Electric Plants, BWR/4," dated March 31, 2004 Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Technical Specification Proposed Change No. 276 Revision to TS Sections 3.3/4.3 for Control Rod System

Dear Sir or Madam:

This letter submits Proposed Technical Specification Change No. 276 for the Vermont Yankee Nuclear Power Station (VY).

Pursuant to 10CFR50.90, VY hereby proposes to amend its Facility Operating License, DPR-28, by incorporating the attached proposed change into the VY Technical Specifications (TS).

The proposed change would revise the surveillance requirements (SR's) for control rod exercising from weekly to monthly in TS 4.3.A.2, revise verification of control rod coupling integrity as described in TS 4.3.B.1, revise the scram insertion time Limiting Conditions for

,Operation (LCO) and SR's as described in TS 3.3.C and 4.3.C, and enhance TS 3.3.D and 4.3.D, the LCO and SR for Control Rod Accumulators.

As applicable, the changes proposed have Specifications (Reference 1) and are discussed submittal. These changes are consistent with the incorporated Improved Standard Technical in the appropriate basis sections within this bases of the ISTS.

Through this change, VY would revise control rod exercising SR from weekly to monthly in accordance'with NUREG-1433, Revision 3 and TSTF-475, Revision 1, revise the control rod coupling integrity SR's by eliminating the surveillances that do not provide positive identification of coupling, enhancing the control rod coupling integrity surveillance test and increasing the frequency in which coupling integrity testing is required. This change also proposes to modify the Scram Insertion Time LCO by establishing a category of "slow" rods. The corresponding Scram Insertion Time SR changes would increase the frequency of scram time testing surveillances in TS; and the testing would be performance based utilizing a representative sample of control rods in accordance with TSTF-460. The proposed changes to the control rod accumulator specifications primarily involve identifying that accumulator operability, and the corresponding SR, are based upon accumulator gas pressure. Corresponding changes to the TS BASES for each of these sections are also provided for information.

TS Bases are controlled under the VY Bases Control Program (TS 6.7.E and 10CFR50.59) and will be revised and issued after approval of these proposed changes.

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 Page 2 Enclosure I to this letter contains supporting information and the safety assessment of the proposed change. contains the determination of no significant hazards consideration. provides the marked-up version of the current TS pages. is the retyped TS pages. is a copy of revised TS Bases pages, provided for information only.

VY has reviewed the proposed Technical Specification change in accordance with 10CFR50.92 and believes that the proposed change does not involve a significant hazards consideration.

VY has also determined that the proposed change satisfies the criteria for a categorical exclusion in accordance with 10CFR51.22(c)(9) and does not require an environmental review.

Therefore, pursuant to 10CFR51.22(b), the preparation of an environmental impact statement or environmental assessment is not warranted.

This letter contains no new regulatory commitments.

Upon acceptance of this proposed change by the NRC, VY requests that the license amendment be issued with a 90 day implementation from its effective date.

If you have any questions or require additional information concerning this submittal, please contact Mr. David J. Mannai at (802) 451-3304.

I declare under ppnalty of perjury, that the foregoing is true and correct.

Executed on pt Sincerely, Ted A. Sullivan Site Vice P"esi nt Vermont Y Tfkee Nuclear Power Station : : : : :

Supporting Information and Safety Assessment (17 pages)

Determination of No Significant Hazards (2 pages)

Marked-up Version of Current TS Pages (12 pages)

Re-typed TS Pages (12 pages)

Revised TS Bases Pages (12 pages) cc list (next page)

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 Page 3 cc list:

Mr. Samuel J. Collins Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr. James S. Kim, Project Manager License Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2A Washington, DC 20555-0001 USNRC Resident Inspector Vermont Yankee Nuclear Power Station 320 Governor Hunt Road P.O. Box 157 Vernon, VT 05354 Mr. David O'Brien, Commissioner Department of Public Service 112 State Street, Drawer 20 Montpelier, VT 05620-2601

Docket No. 50-271 BVY 08-007 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 276 Revision to TS Sections 3.3/4.3 for Control Rod System Supporting Information and Safety Assessment of Proposed Change

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station Page 1 SUPPORTING INFORMATION Purpose The proposed change would revise the surveillance requirements (SR's) for control rod exercising from weekly to monthly in SR 4.3.A.2, revise verification of control rod coupling integrity as described in TS 4.3.B.1, revise the scram insertion time limiting conditions for operation (LCO) and SR's as described in TS 3.3.C and 4.3.C, and enhance TS 3.3.D and 4.3.D, the LCO and SR for Control Rod Accumulators.

Through this change, Vermont Yankee (VY) would revise control rod exercising SR from weekly to monthly in accordance with NUREG-1433, Revision 3 and TSTF-475, Revision 1, revise the control rod coupling integrity SR's by'eliminating the surveillances that do not provide positive identification of coupling, enhancing the control rod coupling integrity surveillance test and increasing the frequency in which coupling integrity testing is required.

This change also proposes to modify the Scram Insertion Time LCO by establishing a category of "slow" rods.

The corresponding Scram Insertion Time SR changes would revise the frequency of scram time testing surveillances; and the testing would be performance-based utilizing a representative sample of control rods in accordance with NRC approved TSTF-460. The proposed changes to the control rod accumulator specifications primarily involve identifying that accumulator operability, and the corresponding SR, is based upon accumulator pressure. Corresponding changes to the TS Bases for each of these sections are also proposed, as appropriate and included for information. Where applicable, the proposed changes are consistent with Standard Technical Specifications 1 (STS), including administrative changes such as capitalizing terms which are defined within the Definitions section of the Technical Specifications.

Background

Control rods are components of the Control Rod Drive (CRD) System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational occurrences, that specified acceptable fuel design limits are not exceeded.

In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase if there were to be a malfunction in the CRD System.

The first part of this proposed Technical Specification change request involves control rod operability and in particular control rod coupling integrity surveillances. The CRD system at VY consists of 89 locking piston control rod drive mechanisms (CRDMs) and a hydraulic control unit for each drive mechanism. The locking piston type CRDM is a double acting hydraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram insertion.

An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet mechanism.

The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.

The second part of this change involves control rod scram time testing. The control rods are scrammed by positive means using hydraulic pressure exerted on the CRD piston. When a scram signal is initiated, control air is vented from the scram valves, allowing them to open by spring action. Opening the exhaust valve reduces the pressure above the main drive piston to atmospheric pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston. Since the notches in the index tube are tapered on the lower edge, 1

NUREG 1433, Revision 3, "Standard Technical Specifications General Electric Plants, BWR/4," dated March 31, 2004

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 2 the collet fingers are forced open by cam action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston.

As the drive moves upward, and the accumulator pressure reduces below the reactor pressure, a ball check valve opens, letting the reactor pressure complete the scram action. If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod in the required time without assistance from reactor pressure.

The third major part of this proposed technical specification change is with regard to the control rod accumulators. The accumulators are part of the CRD System and are provided to ensure that the control rods scram under varying reactor conditions. The accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from the nitrogen overpressure, which provides the required energy.

The scram accumulators are necessary to scram the control rods within the required insertion times of TS 3.3.C "Scram Insertion Times."

An additional change is to revise the frequency of control rod exercising from weekly to monthly in accordance with NUREG-1433, Revision 3 and the recently issued TSTF-475, Revision 1. A reduction in the number of control rod positioning steps will reduce equipment wear and reduce the opportunity for reactivity control errors.

Comparison to Standard Technical Specifications (STS)

In the Preface of NUREG-1433 Revision 3, "Standard Technical Specifications General Electric Plants, BWR/4," states in part:

Licensees are encouraged to upgrade their technical specifications consistent with those criteria and conforming, to the practical extent, to Revision 3 to the improved STS.

This proposed Technical Specification change request is consistent, where applicable, with Standard Technical Specifications (STS) with the recognition that VY has custom Technical Specifications.

The individual proposed changes are discussed in the Safety Assessment portion below with a comparison being drawn between the proposed specifications and STS, as necessary.

STS Surveillance Requirements (SR) for 3.1.3 "Control Rod Operability" are being fully implemented within the VY Technical Specifications. This includes the adoption in total of STS SR 3.1.3.5 regarding the verification that each control rod does not go to the overtravel position.

The proposed change to the control rod scram time LCO and SR's is also consistent with STS.

Of particular note is the adoption of STS LCO 3.1.4 for "Control Rod Scram Times" regarding the limitation of "slow" rods and the corresponding surveillance requirements for STS 3.1.4.

Current Technical Specification (TS) section 3.3.D "Control Rod Accumulators" is proposed to be replaced in its entirety by the adoption of STS Section 3.1.5 "Control Rod Scram Accumulators."

Updated Final Safety Analysis Report (FSAR)

VY FSAR Section 3.4 describes the mechanical aspects of the control rods. The text contains, among other things, an evaluation of the control rods, scram times, analysis of postulated malfunctions related to rod withdrawal, and scram reliability.

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 3 SAFETY ASSESSMENT The proposed change would revise the surveillance requirements (SR's) for verifying control rod coupling integrity as described in TS 4.3.B.1, revise control rod exercising from weekly to monthly in SR 4.3.A.2, revise the scram insertion time limiting conditions for operation (LCO) and SR's as described in TS 3.3.C and 4.3.C, and enhance TS 3.3.D and 4.3.D, the LCO and SR for Control Rod Accumulators.

Through this change, VY would revise the control rod coupling integrity SR's by eliminating the surveillances that do not provide positive identification of coupling, enhancing the control rod coupling integrity surveillance test and increasing the frequency in which coupling integrity testing is required.

This change also proposes to modify the Scram Insertion Time LCO by establishing a category of "slow" rods. The corresponding Scram Insertion Time SR changes would increase the frequency of scram time testing surveillances; and the testing would be performance-based utilizing a representative sample of control rods. The proposed changes to the control rod accumulator specifications primarily involve identifying that accumulator operability, and the corresponding SR, are based upon accumulator pressure. Corresponding changes to the Bases for each of these sections is also proposed, as appropriate.

The proposed changes are consistent with Standard Technical Specifications (STS), where applicable.

Administrative changes are also being proposed by capitalizing terms which are defined within the Definitions section of the technical specifications. This change is consistent with the use of defined terms within STS.

The below Table details each proposed change and provides the basis and safety assessment for each change. It is noted that TS 3.3.C.1.1 & 3.3.C.1.2 are to be revised consistent with the mark-ups identified in Enclosure 3.

Each material change has been identified and justified within the Table below and then, for consistency with STS, the entire text is being replaced by Insert #1. Similar mark-ups and justifications are being made to TS 4.3.C.1 and 4.3.C.2 and then, for consistency with STS, the entire text is being replaced by Inserts #2 (the SR's) and #3 (Table 4.3.C-1).

The same process is followed for TS Section 3.3.D "Control Rod Accumulators" which is being replaced in its entirety by Insert #5.

Change Current Technical Specification Proposed Change Technical Specification (TS) 4.3.B.1 (a)

The subject text of TS 4.3.B.1 (a) would be 1

currently reads 'When a rod is withdrawn deleted based upon the justification the first time subsequent to each refueling provided below.

outage or after maintenance, observe discernable response of the nuclear instrumentation; however, for initial rods when response is not discernable, subsequent exercising of these rods after the reactor is critical shall be performed to verify instrumentation response; and" Basis/Safety Assessment:

The requirement to verify control rod coupling by withdrawing a control rod and observing discernable response of the nuclear instrumentation is deleted. If sufficient friction is present to, uncouple the control rod from its drive, the control rod may not follow the drive being withdrawn.

In this case, a lack of neutron flux level change may be indicative of an uncoupled rod.

However, this is not a positive check that the control rod is uncoupled since, if sufficient friction

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Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 4 is not present, an uncoupled rod could follow the drive being withdrawn. TS SR 4.3.B.1 (b)

(which is being split and renumbered to be 4.3.B.1 (a) and (b)) requires verification that a control rod does not go to the withdrawn over-travel position when fully withdrawn. The over-travel feature provides a positive check of coupling integrity since only an uncoupled control rod can go to the over-travel position.

This change is consistent with STS SR 3.1.3.5 "Control Rod OPERABILITY."

Change Current Technical Specification Proposed Change TS SR 4.3.B.1 (b) currently requires that The subject text of TS 4.3.B.1 (b) would be 2

control rod coupling verification be deleted based upon the justification performed prior to startup following a provided below.

refueling outage by withdrawing each control rod "continuously to observe that the rate of withdrawal is proper."

Basis/Safety Assessment:

TS SR 4.3.B.1 (b) currently requires a control rod coupling verification prior to startup following a refueling outage by withdrawing each control rod continuously to observe that the rate of withdrawal is proper. If sufficient friction is present to uncouple the control rod from its drive, the control rod may not follow the drive being withdrawn and the rate of control rod drive withdrawal may be slower than normal. However, this is not a positive check that the control rod is uncoupled since, if sufficient friction is not present, an uncoupled rod could follow the drive being withdrawn and the rate of withdrawal may not be affected. The proposed revision to SR 4.3.B.1(b) (which is consistent with STS SR 3.1.3.5) requires verification that a control rod does not go to the withdrawn over-travel position. The over-travel feature provides a positive check of coupling integrity since only an uncoupled control rod can go to the over-travel position when fully withdrawn. This verification is required to be performed in STARTUP and RUN MODEs any time a control rod is withdrawn to the full out position and prior to declaring a control rod OPERABLE after work on the control rod or Control Rod Drive system that could affect coupling. As a result, SR 4.3.B.l(b) (which is being split and renumbered to be 4.3.B.1(a) and 4.3.B. 1 (b)) provides adequate assurance that the control rods are coupled.

This change is consistent with STS SR 3.1.3.5 "Control Rod OPERABILITY."

Change Current Technical Specification Proposed Change TS SR 4.3.B.1 (b) currently requires that The subject text of TS 4.3.B.1 (b) would be 3

each control rod be verified as coupled revised based upon the justification prior to "startup following a refueling provided below.

outage" and "following uncoupling." In addition, it also requires that following uncoupling, each drive and blade be coupled and fully withdrawn to verify positive coupling.

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 5 Basis/Safety Assessment:

The TS SR 4.3.B.1 (b) requirement to verify that each control rod is coupled prior to startup following a refueling outage and following uncoupling are deleted. In addition, TS 4.3.B.1(b) requirement to couple the drive and blade and fully withdraw it is implicit in the coupling verification requirement of TS SR 4.3.B.1(b) and is also deleted. The proposed revision toTS SR 4.3.B.1 (b) (which is the same as STS SR 3.1.3.5) requires verification that a control rod does not go to the withdrawn over-travel position. The over-travel feature provides a positive check of coupling integrity since only an uncoupled control rod can go to the over-travel position. This verification is required to be performed in STARTUP and RUN MODEs any time a control rod is withdrawn to the full-out position and prior to declaring a control rod OPERABLE after work on the control rod or Control Rod Drive system that could affect coupling.

As a result, SR 4.3.B.1(b) (which is being split and renumbered to be SR 4.3.B.1(a) and (b))

provides adequate assurance that the control rods are coupled.

This chanqe is consistent with STS SR 3.1.3.5 "Control Rod OPERABILITY."

Change Current Technical Specification Proposed Change TS SR 4.3.B.1(b) currently requires for The subject text of TS 4.3.B.1(b) and 4

coupling tests that: "the results of each 4.3.B.2 would be deleted based upon the test be recorded."

justification provided below.

TS SR 4.3.B.2 for CRD Housing Support System inspections requires that: "the results of the inspection recorded."

Basis/Safety Assessment:

The TS SR 4.3.B.1(b) provides details regarding record keeping. These details are not necessary to ensure the associated CRD and control rod blade are coupled and are to be deleted. The proposed revision to TS SR 4.3.B.1(b) (which is based on STS SR 3.1.3.5) requires the same testing without going into record keeping details and demonstrates the same OPERABILITY; these requirements are adequate for ensuring each associated CRD and control rod blade are coupled.

TS SR 4.3.B.2 also provides details regarding record keeping. These details are not necessary to be included in the TS to ensure the CRD Housing Support System is configured properly and are to be deleted.

The requirement for retention of records related to activities affecting quality is contained in 10CFR50, Appendix B, Criterion XVII which VY already applies to TS surveillances. These record retention requirements provide a record of certain activities important to plant safety, but the records themselves do not assure safe operation of the facility. As such, the record retention details do not need to be duplicated in the TS to provide adequate protection of the public health and safety.

This chanae is consistent with STS SR 3.1.3.5 "Control Rod OPERABILITY."

Change Current Technical Specification Proposed Change TS 4.3.B.1(b) currently reads: 'The The subject text of TS 4.3.B.1(b) would 5

position and over-travel lights shall be be deleted based upon the justification observed."

provided below.

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 6 Basis/Safety Assessment:

The TS 4.3.B.1 (b) details of how to determine if a control rod has reached the over-travel position are relocated to the Bases for TS 4.3.B.1. These details are not necessary to ensure the associated CRD and control rod blade are coupled. TS SR 4.3.B.1(b) requirements (which are being split and renumbered to be 4.3.B.1 (a) and (b)) for verifying each control rod does not go to the withdrawn over-travel position are adequate for ensuring the associated CRD and control rod blade are coupled. As such, these relocated details are not required to be in the TS to provide adequate protection of the public health and safety. Changes to the Bases are controlled by the provisions of the VY Bases Control Program as described in Chapter 6 of the VY TS.

This chanqe is consistent with STS SR 3.1.3.5 "Control Rod OPERABILITY."

Change Current Technical Specification Proposed Change TS 3.3.C.1.1 & 3.3.C.1.2 currently require TS 3.3.C.1.1 & 3.3.C.1.2 are to be revised 6

that average control rod scram insertion consistent with the mark-ups identified times be within certain limits and also (with each material change identified and contain the control rod scram time justified) and then replaced in total by A surveillance acceptance criteria.

Inserts #1, 2 & 3.

Basis/Safety Assessment:

The purpose of the control rod scram time LCO is to ensure the negative scram reactivity insertion rate corresponding to that used in licensing basis calculations is supported by individual control rod drive scram performance distributions allowed by the Technical Specifications. Current TS 3.3.C.1.1 and 3.3.C.1.2 accomplish the above purpose by placing requirements on maximum individual control rod drive scram times (7.00 second requirement),

average scram times and local scram times (average of three fastest control rods in 2x2 arrays).

The scram time requirements in section 3.3.C.1.1 of current Technical Specifications have no associated action statement. These scram time requirements applied at VYNPS in the past when faster than "standard" analytical scram times were used to justify faster scram time performance in transient analysis [the so-called Measured Scram Time option in GESTAR].

Several years ago, in place of Measured Scram Times, VYNPS adopted a similar margin improvement strategy, ODYN Option B [see GESTAR]. The scram time requirements for ODYN Option B are in the Core Operating Limits Report (COLR). Section 3.3.C.1.2 scram time requirements are equivalent to the standard analytical times used in transient analyses.

Proposed TS 3.3.C.1 provides a different method to determine if measured scram insertion times are sufficient to insert the amount of negative reactivity assumed in the accident and transient analyses from current TS 3.3.C.1.1 and 3.3.C.1.2. A description and supporting analysis for the proposed TS 3.3.C.1 method (which is identical to that utilized by NUREG-1433, Standard Technical Specification LCO 3.1.4) is contained in BWROG-8754, letter from R.F.

Janecek (BWROG) to R.W. Starostecki (NRC), dated September 17, 1987.

The methodology used in the design basis transient analysis (one dimensional neutronics),

assumes that all control rods scram at the same speed. This is called the analytical scram time requirement. Performing an evaluation assuming all control rods scram at the analytical limit will result in the generation of a scram reactivity versus time curve that is called the analytical scram reactivity curve. It is the purpose of the scram time LCO to ensure that, under allowed plant conditions, this analytical scram reactivity will be met. Since scram reactivity cannot be readily measured at the plant, the safety analyses use appropriately conservative scram reactivity versus insertion fraction curves to account for the variation in scram reactivity during a cycle.

Therefore, the Technical Specifications must only ensure the scram times are satisfied.

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 7 If all control rods scram at least as fast as the analytical limit, the analytical scram reactivity curve will be met. However, it is also known that a distribution of scram times (some slower and some faster than the analytical limit) can also provide adequate scram reactivity. By definition, for a situation where all control rods do not satisfy the analytical scram time limits, the condition is acceptable if the resulting scram reactivity meets or exceeds the analytical scram reactivity curve. This can be evaluated using models which allow for a distribution of scram speeds. It follows that the more control rods that scram slower than the analytical limit, the faster the remaining control rods must scram to compensate for the reduced scram reactivity insertion rate of the slower control rods. Proposed TS 3.3.C.1 incorporates this philosophy by specifying scram time limits for each individual control rod instead of specifying limits on the average of all control rods or the average of 2x2 arrays. This philosophy is similar to that currently being used for BWR/4 plants that have converted to Improved Technical Specifications. Proposed TS 3.3.C.1 scram time limits provide margin to the analytical scram time limits to allow for a specified number and distribution of slow control rods.

The scram times specified in proposed Table 4.3.C-1, reflect time measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes

("pickup") when the index tube passes a specific location and then opens ("dropout") as the index tube travels upward. Verification of the specified scram times in proposed Table 4.3.C-1 is accomplished through the measurement of "dropout" times. The scram insertion times listed in proposed Table 4.3.C-1 reflect the required scram insertion time Values listed in Table 3.4 from'the BWROG-8754 Report. Therefore, if all control rods meet the proposed LCO scram time limits found in proposed Table 4.3.C-1, as measured from the de-energization of scram pilot valve solenoids at time zero (Note a), the analytical scram reactivity assumptions are satisfied. If any control rods do not meet the LCO time limit, the LCO specifies the number and distribution of these "slow" control rods to ensure the analytical scram reactivity assumptions are still satisfied.

If the number of "slow" rods is excessive (> 7% of 89 or > 6) or do not~meet the distribution requirements, the unit must be shutdown. This change is considered more restrictive on plant operation since the proposed individual times, as a whole, are more restrictive from the average times. That is, currently, the "average time" of all rods or a group of rods can be improved by a few fast scramming rods, even when there may be more than 6 "slow" rods, as defined in the proposed specification. Therefore, the proposed specification limits the number of slow rods to 6, imposes individual (vs. average) scram times and ensures no more than 2 OPERABLE control rods that are "slow" occupy diagonally or face adjacent locations.

The Table 4.3.C-1 Control Rod Scram Times are the 'last" times from BWROG-8754 Table 3-4.

Except for notch 46, these times are slightly faster than the analytical ("average") times in the current section 3.3.C to compensate for the presence of a small number of "slow" control rods.

The notch 46 (-5%) insertion time is typically the most limiting requirement for CRD performance, however, this requirement has been shown to have negligible impact on plant transient performance. Actual drive performance is significantly better than the 'last' limits as seen from the following table of scram time tests performed at the start of the current operating cycle.

Notch Position Average of all rods Slowest Rod Proposed Table (seconds)

(seconds) 4.3.C-1 (seconds) 46 0.294 0.326 0.44 36 0.804 0.865 1.08 26 1.328 1.447 1.83 06 2.440 2.654 3.35

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 8 The notch 46 scram time requirement applied for, is consistent with the analytical scram time assumption for VYNPS.

This change is more restrictive in the sense that each control rod, with the exception of "slow" rods, must fully insert faster than the current average scram times requirement of current TS Table 3.3.C.1.2.

The current maximum scram time requirement of TS 3.3.C.2 has been retained for the purpose of defining the threshold between a "slow" control rod and an inoperable control rod. The proposed Note to Table 4.3.C-1 (Note 2) ensures that a control rod is not inadvertently considered "slow" when the scram time exceeds 7 seconds.

This change is consistent with STS 3.1.4 "Control Rod Scram Times."

Change Current Technical Specification Proposed Change VY's current Technical Specifications The proposed change would expand the 7

require that the control rod scram times applicability of control rod scram times to satisfy the requirements of the tables in be in STARTUP or RUN MODEs.

TS 3.3.C.1 during "...the reactor power operation condition..."

Basis/Safety Assessment:

TS 3.3.C.1.1 and 3.3.C.1.2 established the Applicability of minimum scram times as "in the reactor power operation condition." The proposed TS 3.3.C.1 has minimum scram time limits applicable during the STARTUP and RUN MODEs. This change is more restrictive than the existing requirement because it would apply to all conditions where a reactor scram may be required by the accident analysis, including reactor startup and power ascension.

This change is consistent with STS 3.1.4 "Control Rod Scram Times."

Change Current Technical Specification Proposed Change In TS 3.3.C.1 the scram insertion time, The subject column of TS 3.3.C.1 would 8

limits are specified in terms of "% Inserted be deleted based upon the justification From Fully Withdrawn."

provided below.

Basis/Safety Assessment:

TS 3.3.C.1.1 and 3.3.C.1.2 scram insertion time limits are specified in terms of "% Inserted From Fully Withdrawn." Scram times are measured from signals generated by reed switches corresponding to control rod notch positions. The proposed TS 3.3.C.1 would specify the scram insertion time limits in terms of "notch position" within a specified number of seconds.

This will eliminate the need to convert notch position to "% Inserted From Fully Withdrawn" to verify acceptance criteria. Since the only effect of specifying limits in terms of notch position instead of "% Inserted From Fully Withdrawn" is to eliminate the need to convert the units after performance of a test, this is an administrative change.

This change is consistent with STS 3.1.4 "Control Rod Scram Times."

mnt Technical Specification Proposed Change

-I TS 3.3.C.2 states 'The maximum scram insertion time for 90% insertion of any ooerable control rod shall not exceed 7.00 The subject text of TS 3.3.C.2 would be revised to state "The maximum scram insertion time to notch position 06 of any

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Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 9 seconds." TS currently do not contain a operable control rod shall not exceed surveillance requirement (SR) to verify 7.00 seconds." In addition, TS SR the subject LCO.

4.3.C.2 is being proposed (reference Insert # 4) to verify the subject maximum scram times.

Basis/Safety Assessment:

TS 3.3.C.2 requires the maximum scram insertion time for 90% insertion of any operable control rod to not exceed 7.00 seconds. It is proposed that TS 3.3.C.2 require that each operable control rod have a maximum scram time from fully withdrawn to notch position 06 be

< 7.00 seconds. In addition, TS SR 4.3.C.2 is being proposed (reference Insert # 4) to perform a verification of the above maximum scram time.

Redefining the 90% insertion to a notch position and adding a SR to perform a verification of the LCO does not eliminate any of the existing requirements or impose a new or different treatment of the requirement. The 90% insertion is converted to notch position 06 which is approximately 88% insertion from fully withdrawn. Specifying notch position 06 is the typical industry practice, as specified in BWR/4 Standard Technical Specifications. Currently, the insertion time for 90% insertion is a calculated interpolation, and the proposed new requirement would be based upon an actual control rod notch position measurement., Notch position 06 is approximately the same as the 90% insertion distance and will now be based upon actual times to a notch position of a control rod and eliminate calculating and interpolating of control rod scram performance.

This change is consistent with STS 3.1.3 "Control Rod Operability."

Change Current Technical Specification Proposed Change I

TS 4.3.C.1 requires each control rod to be The proposed text of TS 4.3.C.1 would 10 scram time tested with reactor steam require each control rod to be scram time dome pressure > 800 psig prior to tested with reactor steam dome pressure exceeding 30% RATED THERMAL

> 800 psig prior to exceeding 30% RTP POWER (RTP) after each refueling after each reactor shutdown of > 120 outage.

days.

Basis/Safety Assessment:

TS 4.3.C.1 requires each control rod to be scram time tested with reactorsteam dome pressure > 800 psig prior to exceeding 30% RTP after each refueling outage. The proposed TS 4.3,C.1.a would require each control rod to be scram time tested with reactor steam dome pressure > 800 psig prior to exceeding 30% RTP after each reactor shutdown > 120 days. To ensure that scram time testing is performed within a reasonable time following a refueling or after a shutdown duration >120 days or longer, control rods are required to be tested before exceeding 30% RTP following the shutdown. As such, this is an additional restriction on plant operation which constitutes a more restrictive change.

Proposed SR 4.3.C.1.d will require scram time testing after fuel movement in an affected core cell prior to exceeding 30% RTP. In a typical refueling outage, this SR will require all control rods to be tested before exceeding 30% RTP following the shutdown.

This change is consistent with STS 3.1.4 "Control Rod Scram Times."

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 10 Change Current Technical Specification Proposed Change TS 4.3.C.2 requires control rod scram The proposed text of TS 4.3.C.1 would 11 time testing during or following a require a verification, for a representative controlled shutdown of the reactor, but sample, that each tested control rod not more frequently than 16 weeks nor scram time is within limits with reactor less frequently than 32 week [224 days]

steam dome pressure > 800 psig each intervals for 50% of the control rod drives 200 days cumulative operation in the in each quadrant of the reactor core.

RUN MODE.

Basis/Safety Assessment:

TS 4.3.C.2 requires control rod scram time testing during or following a controlled shutdown of the reactor, but not more frequently than 16 weeks nor less frequently than 32 week intervals

[224 days] for 50% of the control rod drives in each quadrant of the reactor core. The proposed text of TS 4.3.C.1 would require a verification, for a representative sample, that each tested control rod scram time is within limits with reactor steam dome pressure Ž 800 psig each 200 days cumulative operation in RUN MODE. The 200 day Frequency is based on industry operating experience that has shown control rod scram times do not significantly change over an operating cycle. This increased surveillance frequency is an additional restriction on plant operation which constitutes a more restrictive change.

This change is consistent with a notice announcing the availability of a similar proposed TS change using the consolidated line item improvement process (CLIIP) which was published in the Federal Register on August 23, 2004 (69 FR 51854). These changes are based on TS Task Force (TSTF) change traveler TSTF-460 (Revision 0) that has been approved generically for the boiling water reactor (BWR) Standard TS, NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) by revising the frequency of SR 3.1.4.2, control rod scram time testing,'from '120 days cumulative operation in MODE 1 [RUN MODE]" to "200 days cumulative operation in MODE 1."

VY has reviewed the safety evaluation (SE) published on August 23, 2004 (69 FR 51854) as part of the CLIIP Notice of Availability. This verification included a review of the NRC staff's SE and the supporting information provided to support TSTF-460. VY has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to VY and justify this amendment for incorporation to the VY TS.

As described in the CLIIP model SE, part of the justification for the change in surveillance frequency is the high reliability of the control rod drive system. As requested in the notice of availability published on August 23, 2004 (69 FR 51854), the historical performance of the control rod drive system at VY is as follows:

Over a period of approximately the last 10 years, there have been over 1700 scram time tests conducted. During this period, none of the rods had scram times that would have required the rods to be declared "slow" and none of the rods were determined to be "inoperable."

It is noted that the corresponding BASES for this proposed change would be revised to reflect that the control rod insertion time acceptance criterion for the percentage of slow rods allowed, would be 7.5 percent of the random at-power surveillance sample (with the surveillance period of 200 cumulative days of operation in RUN MODE). The more restrictive 7.5 percent acceptance criterion for testing the random sample is consistent with the TS 4.3.C.2 (STS TS 3.1.4) objective of ensuring that no more than 6 OPERABLE control rods are slow at any given time.

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 11 Change Current Technical Specification Proposed Change TS 4.3.C.2 currently states "50% control The corresponding proposed change to 12 rod drives in each quadrant of the reactor TS 4.3.C.2 (to be renumbered as SR core shall be measured for scram times 4.3.C.1) would require: Every 200 days specified in Specification 3.3.C. All cumulative operation in RUN MODE, control rod drives shall have experienced verify, for a representative sample, that scram-time measurements each year."

each tested control rod scram time is within the limits of Table-4.3.C-1 with reactor steam dome pressure > 800 psig.

Basis/Safety Assessment:

TS 4.3.C.2 requires control rod scram time testing during or following a controlled shutdown of the reactor, but not more frequently than 16 weeks nor less frequently than 32 week intervals for 50% of the control rod drives in each quadrant of the reactor core. Proposed TS SR 4.3.C.1 would require: Every 200 days cumulative operation in RUN MODE, verify, for a representative sample, that each tested control rod scram time is within the limits of Table 4.3.C-1 with reactor steam dome pressure > 800 psig. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow." With more than 7.5% of the sample declared to be "slow" per the criteria in TS SR 4.3.C.1, Table 4.3.C-1, additional control rods are tested until this 7.5% criterion (i.e., 7.5% of the entire sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all surveillances) exceeds the LCO limit.

The proposed change is less restrictive since the number of control rods tested during each control rod scram time test is reduced from 50% to 10% and the total amount tested in a calendar year is reduced from 100% to the 10%-20% range. While the total number of control rods scram time tested will be reduced, the frequency of the testing will be increased (see change # 11) from as much as once every 224 days to once every 200 days. The Surveillance Requirement will still be performed at a frequency that industry operating experience has shown to be adequate for maintaining control rod scram times within required limits. The 200 day Frequency and the number of control rods tested is based on industry and VY operating experience that has shown control rod scram times do not significantly change over an operating cycle.

This change is consistent with a notice announcing the availability of a similar proposed TS change using the CLIIP that was published in the Federal Register on August 23, 2004 (69 FR 51854). These changes are based on TS Task Force (TSTF) change traveler TSTF-460 (Revision 0) that has been approved generically for the boiling water reactor (BWR) Standard TS, NUREG-1433 (BWR/4) and NUREG-1434 (BWR/6) by revising the frequency of SR 3.1.4.2, control rod scram time testing, from "120 days cumulative operation in MODE 1" to "200 days cumulative operation in MODE 1." Please reference the expanded discussion and justification contained within proposed change #11 above.

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 12 Change Current Technical Specification Proposed Change TS 4.3.C:2 currently states 'Whenever The subject text of TS 4.3.C.2 would be 13 50% of the control rod drives scram times deleted based upon the justification have been measured, an evaluation shall provided below.

be made to provide reasonable assurance that proper control rod drives performance is being maintained. The results of measurements performed on the control rod drives shall be submitted in the start up test report."

Basis/Safety Assessment:

It is proposed that the TS 4.3.C.2 details concerning the evaluation of control rod performance be deleted. The records associated with the performance of Technical Specification required Surveillances are required to be maintained as part of our Quality Assurance Program.

Changes to the Quality Assurance Program are controlled by 1 OCFR50.54(a). Furthermore, evaluation of TS surveillance performance and acceptance criteria is part of our surveillance procedures and would not typically be reported to NRC in a submittal. Specifying these details in the Technical Specifications is not necessary to ensure control rod scram times are within limits. Proposed SR 4.3.C.2 (to be renumbered as SR 4.3.C.1) and associated Table 4.3.C-1 are adequate to ensure scram time testing is performed and the scram times are within acceptable limits. As such, these details are not required to be in the Technical Specifications to provide adequate protection of the public health and safety.

This change is consistent with STS 3.1.4 "Control Rod Scram Times."

Change Current Technical Specification Proposed Change TS 3.3.C.3 currently specifies "If The subject text would be revised to read 14 Specification 3.3.C.1.2 cannot be met, the "If Specification 3.3.C.1 cannot be met, reactor shall not be made supercritical; if the reactor shall not be made operating, the reactor shall be shut down supercritical; if operating, the reactor shall immediately upon determination that be placed in the HOT SHUTDOWN average scram time is deficient."

condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ."

Basis/Safety Assessment:

TS 3.3.C.3 would be revised to identify that Specification 3.3.C.1.2 has been renumbered to be 3.3.C.1. In addition, the proposed change would define the action "immediately" as "12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months " and would define the condition "shut down" as HOT SHUTDOWN.

Since the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analysis when control rod scram time requirements in TS 3.3.C.1 are not met, placing the unit in the HOT SHUTDOWN condition ensures that the unit is brought into a condition where TS 3.3.C.1 does not apply. Cooling down the unit does not provide any additional margin and, in some cases, could be counterproductive since positive reactivity is inserted during a cool-down. Given that the only difference between HOT SHUTDOWN and COLD SHUTDOWN is the temperature requirement, this proposed change is administrative and provides clarity to operator actions.

This change is consistent with STS 3.1.4 "Control Rod Scram Times."

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 13 Change Current Technical Specification Proposed Change TS 3.3.C.4 currently reads "If The subject text of TS 3.3.0.4 would be 15 Specification 3.3.0.2 cannot be met, the revised to specify that the deficient control deficient control rod shall be considered rod be considered inoperable, fully inoperable, fully inserted into the core and inserted and disarmed. The details of the electrically disarmed."

method used to disarm will not be discussed.

Similar wording (i.e.; electrically disarmed) is utilized in TS 3.3.A.2 and TS A similar change is proposed for TS 3.3.B.1.

sections 3.3.A.2, 3.3.B.1 and 3.12.D.2 to delete the word "electrically."

Basis/Safety Assessment:

The TS 3.3.0.4 details of the methods for disarming control rod drives are proposed to be relocated to the Bases. These details are not necessary to ensure the associated ORDs of inoperable control rods are disarmed. Proposed TS 3.3.0.4, which requires disarming the associated ORDs of inoperable control rods, is adequate for ensuring associated CRDs and inoperable control rods are disarmed. As such, these relocated details are not required to be in the Technical Specifications to provide adequate protection of the public health and safety.

Changes to the Bases are controlled by the provisions of the Bases Control Program described in Section 6.7.E of the VY Technical Specifications.

A similar change is proposed for TS 3.3.A.2, 3.3.B.1 and 3.12.D.2 to delete the word "electrically."

This change, is administrative in that the Technical Specifications will still require the subject control rods to be disarmed. The proposed change would allow for the disarming to be either hydraulic or electrical and the method does not need to be prescribed in the Technical Specifications. Since either method provides adequate protection, the change removes

'unnecessary detail and is considered administrative.

This change is consistent with STS 3.1.3 "Control Rod Operability."

Change Current Technical Specification Proposed Change TS 3.3.C.4 currently reads "If The subject text of TS 3.3.0.4 would be 16 Specification 3.3.C.2 cannot be 'met, the enhanced by prescriptive wording that deficient control rod shall be considered would require the insertion of an inoperable, fully inserted into the core and inoperable control rod within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and electrically disarmed."

to have the control rod disarmed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

However, the LCO fails to specify the time frames in which these actions are required to be completed.

Basis/Safety Assessment:

TS 3.3.C.2 identifies that the maximum scram time for any operable control rod shall not exceed 7.00 seconds.

Entergy Nuclear Operations, Inc.

Letter ýumber: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 14 The current TS 3.3.C.4 requires any control rod which can not satisfy TS 3.3.C.2 to be considered inoperable, fully inserted into the core, and disarmed. The proposed change would require that if a control rod can not satisfy TS 3.3.C.2, the subject rod is to be declared inoperable and then fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The allowed completion times are reasonable, considering the small number of allowed inoperable control rods, and to provide time to insert and disarm the control rod in an orderly manner, without challenging plant systems. These times are also consistent with Standard Technical Specifications. As such, this is an additional restriction on plant operation which constitutes a more restrictive change.

This change is consistent with STS 3.1.3 "Control Rod Operability."

Change Current Technical Specification Proposed Change TS 3.3.D provides conditions associated It is proposed that TS 3.3.D would be 17 with when a control rod accumulator may revised and replaced in its entirety with be inoperable.

Insert #5.

Basis/Safety Assessment:

TS 3.3.D provides conditions associated with when a control rod accumulator may be inoperable. The proposed TS 3.3.D would require each control rod scram accumulator to be OPERABLE in STARTUP and RUN MODEs. The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure. In STARTUP and RUN MODEs, the scram function is required for mitigation of Design Basis Accidents (DBAs) and transients, and therefore the scram accumulators must be OPERABLE to support the scram function.

Proposed TS 3.3.D.1, 3.3.D.2 and 3.3.D.3, allow up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depending upon the number of inoperable accumulators and the reactor pressure, before the control rod associated with the inoperable accumulator must be declared inoperable.

Proposed TS 3.3.D.1 would allow for one control rod scram accumulator to be inoperable for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , provided the reactor pressure is > 800 psig (pressure based upon current TS BASES). As noted in Section 3.4.5.4 of the VY Updated Final Safety Analysis Report (UFSAR): "Each drive requires about 2.5 gallons of water during the scram stroke. There is adequate water capacity in each drive's accumulator to complete a scram in the required time at low reactor vessel pressure. At higher reactor vessel pressures, the accumulator is assisted on the upper end of the stroke by reactor vessel pressure acting on the drive via the ball check (shuttle) valve. As water is forced from the accumulator, the accumulator discharge pressure falls below reactor vessel pressure. This causes the check valve to shift its position to admit reactor pressure under the drive piston. Thus, reactor vessel pressure furnishes the force needed to complete the scram stroke at higher reactor vessel pressures. When the reactor vessel is up to full operating pressure, the accumulator is actually not needed to meet scram time requirements." The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on the large number of control rods available to provide the scram function and the ability of the affected control rod to scram only with reactor pressure at high reactor pressures. In addition, proposed'TS 3.3.D.1.a provides an option to declare a control rod with an inoperable scram accumulator "slow." Action to declare the control rod "slow" allows the rod to remain withdrawn but not disarmed.

Entergy.Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 15 Disarming the inoperable rod is intended to prevent inadvertent operation. The limits and allowances for numbers and distribution of inoperable and "slow" control rods (found in TS 3.3.A.2 and 3.3.C.1 respectively) are appropriately applied to control rods with inoperable scram accumulators whether declared inoperable or "slow." The option for declaring the control rod with an inoperable accumulator "slow" is restricted (by a Note to 3.3.D.1.a and 3.3.D.2.b.1) to control rods that were not previously known to be "slow." This restriction prevents allowing a "slow" control rod from remaining OPERABLE with the additional degradation to scram time caused by an inoperable scram accumulator.

Proposed TS 3.3.D.2 allows two or more control rod scram accumulators to be inoperable for, up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months when reactor pressure is > 800 psig. The requirement for declaration of "slow" or inoperable (and the implied concurrent restoration allowed time) is provided in proposed TS 3.3.D.2.b.1 and b.2. This 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance provides a reasonable time to attempt investigation and restoration of the inoperable accumulator. The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable.

Furthermore, proposed TS 3.3.D.2.a addresses the situation where additional accumulators may be rapidly becoming inoperable due to loss of charging water header pressure. Once verification of adequate charging water header pressure is made (20 minutes is provided), and considering that reactor pressure is adequate to assure the scram function of the control rods with inoperable accumulators, the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months extension is not significant.

Proposed TS 3.3.D.3 allows one or more control rod scram accumulators to be inoperable for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months when reactor pressure is < 800 psig. This 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance provides a reasonable time to attempt investigation and restoration of the inoperable accumulators.

Proposed 3.3.D.3.a addresses the situation where additional accumulators may berapidly becoming inoperable due to a loss of charging water header pressure. The verification is similar to that described in proposed TS 3.3.D.2.a above; however, the verification must be made immediately since adequate scram pressure is not guaranteed without the CRD system in operation. Once verification of adequate charging water header pressure is made, the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months extension is not significant. In addition, since the reactor pressure may not be adequate to scram the rods in a proper time, the allowance provided in proposed TS 3.3.D.1 and 2 (to declare the rod "slow") is not provided under the lower pressure condition (<800 psig).

Proposed TS 3.3.D.4 provides the'required actions if the CRD system verification-is not satisfactory. If the system pressure is not adequate, it is necessary to fully, insert all control rods within one hour. This ensures that the extensions of proposed TS 3.3.D.2 and 3 will not be used unless adequate CRD pressure is available to scram the reactor.

Proposed TS 3.3.D includes a Note ("Separate action item entry is allowed for each control rod scram accumulator") which provides more explicit instructions for proper. application of the required actions to ensure Technical Specification compliance. This Note provides direction consistent with the intent of the existing actions for inoperable control rod scram accumulators.

Upon discovery of each inoperable accumulator, it is intended that, each specified action be applied regardless of it having been applied previously for other inoperable accumulators.

This change is consistent with STS 3.1.5 "Control Rod Scram Accumulators" except for proposed TS 3.3.D.4 which is proposing a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time vs. an immediate Completion Time as in STS. However,.this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months remains acceptable and is significantly more restrictive than the current TS requirement to be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 16 Change Current Technical Specification Proposed Change TS 4.3.D currently requires a check of the The subject text would be revised to 18 status of the pressure and level alarms for require that each control rod scram each accumulator once per shift.

accumulator nitrogen pressure be verified I_

to be -> 940 psig every 7 days.

Basis/Safety Assessment:

Technical Specification SR 4.3.D currently requires a check of the status of the nitrogen pressure and level alarms for each accumulator once each shift. It is proposed to modify SR 4.3.D to be consistent with STS SR 3.1.5.1 which requires that each control rod scram accumulator pressure be verified to be -> 940 psig every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. Verifying level does not assure OPERABILITY. The change in surveillance frequency from once per shift to once per 7 days has been shown to be acceptable through industry operating experience and takes into account indications and alarms available in the Control Room.

This change is consistent with STS 3.1.5 "Control Rod Scram Accumulators."

Change Current Technical Specification Proposed Change TS 3.3.F currently requires that the plant TS 3.3.F would be deleted and a 19 be placed in the cold shutdown condition corresponding shutdown action statement within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if Specifications 3.3.B would be added as 3.3.B.6.

I through 3.3.D are not satisfied.

Basis/Safety Assessment:

TS 3.3.F currently requires that the plant be placed in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if Specifications 3.3.B through 3.3-D are not satisfied. Since TS 3.3.C and 3.3.D will have their own specific action statements, this proposed change would relocate the existing shutdown action statement from 3.3.F to 3.3.B.6 and would provide Required Actions if specifications 3.3.B.1 through 3.3.B.5 are not satisfied. Existing TS 3.3.F is not warranted for TS 3.3.0 since it already contains specific action statements. Existing TS 3.3.F is also not warranted for TS 3.3.D due to the addition of specific action statements as proposed in Changes #16, 17 and 18 above.

Change Current Technical Specification Proposed Change TS 3.3.B.1 currently states that each TS 3.3.B.1 is proposed to be revised to 20 control rod shall be coupled to its drive or remove the wording pertaining to number inserted in the core. It further states that of control rods per quadrant.

when removing up to one control rod per quadrant for inspection and the reactor is in the refueling mode, the coupling requirement does not apply.

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 17 Basis/Safety Assessment:

TS 3.3.B.1 currently requires that each control rod shall either be coupled or inserted and disarmed. The specification also allows relaxation of this specification for removing one control rod drive per quadrant for inspection when in the refueling mode. In current TS 3.12 "Refueling and Fuel Handling," TS 3.12.D and 3.12.E contain guidance for removal of one or more control rods and/or control rod drives for inspection/maintenance. TS 3.12 contains the requirements for control rods in the refueling mode and thus the wording in TS 3.3.B.1 with regard to numbers and locations of control rod(s) that can be uncoupled during refueling is both unnecessary and incorrect. This "up to one per quadrant' wording is proposed to be deleted from this specification, as control rod and control rod drive inspection/maintenance requirements are already addressed in TS 3.12.

Change Current Technical Specification Proposed Change TS 4.3.A.2 currently states: "Each TS 4.3.A.2 would be revised from a 21 partially or fully withdrawn operable weekly to monthly surveillance.

control rod shall be exercised one notch at least once each week."

Basis/Safety Assessment:

TS 4.3.A.2 currently requires that the control rods be exercised one notch each week. The revised surveillance would change the exercising to monthly.

This change is based upon STS and TSTF-475, Revision1. STS currently allows partially withdrawn control rods to be exercised one notch every 31 days. TSTF-475, Revision 1 allows for fully withdrawn control rods to also be exercised at the same monthly frequency.

The purpose of this surveillance is to confirm control rod insertion capability, and, based upon industry operating experience, a stuck control rod is an extremely rare event. This change from weekly to monthly frequency has been determined to be acceptable'because of the demonstrated historic reliability of the CRDMs and that monthly testing will still provide a large number of tests to provide confidence that any problems with the system would be identified.

The reduction in the number of control rod positioning steps prevents unnecessary control rod manipulations and has a two fold benefit. First, it will reduce the duty on the Reactor Manual Control System and CRD hardware, which will improve equipment reliability because it reduces the number of control rod manipulations. Second, it reduces the number of potential reactivity control errors that could occur, because it reduces the number of operator actions:

The potential effects of reducing the number of notch tests are far outweighed by the benefits of (1) reducing undue equipment wear, (2) reducing unnecessary burden on reactor operators and (3) reducing the potential for mispositioning events which accompanies any control rod manipulation.

-This change in TS surveillance frequency is consistent with STS SR 3.1.3.3 and TSTF-475, Revision 1.

Docket No. 50-271 BVY 08-007 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 276 Revision to TS Sections 3.3/4.3 for Control Rod System Determination of No Significant Hazards Consideration

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 1 Determination of No Significant Hazards Consideration Description of amendment request:

The proposed change would substantially revise the surveillance requirements (SR's) for verifying control rod coupling integrity as described in Technical Specification (TS) 4.3.B.1, revise control rod exercising frequency from weekly to monthly in TS 4.3.A.2, revise the scram insertion time Limiting Condition for Operation (LCO) and SR's as described in TS 3.3.C and 4.3.C, and enhance TS 3.3.D and 4.3.D, the LCO and SR for Control Rod Accumulators. Minor revisions and enhancements to TS 3.3.A, 3.3.B and 3.3.C are also included.

Through this change, Vermont Yankee Nuclear Power Station (Vermont Yankee) would revise the control rod coupling integrity SR's by eliminating the surveillances that do not provide positive identification-of coupling and increasing the frequency in which coupling integrity testing is required. The frequency of exercising control rods is proposed to be revised from weekly to monthly. This change also proposes to modify the Scram Insertion Time LCO by establishing a category of "sl6w" rods. The corresponding Scram Insertion Time SR changes would increase the frequency of scram time testing surveillances; and the testing would be performance based utilizing a representative sample of control rods.. The proposed changes to the control rod accumulator specifications primarily involve identifying that accumulator operability, and the corresponding SR, is based upon accumulator pressure. Corresponding changes to the BASES for each of these sections is also provided for information, as appropriate. All of the proposed changes are consistent with Standard Technical Specifications 2 (STS); including administrative changes associated with usage rules, content and format, such as capitalizing terms which are defined within the Definitions section of the Technical Specifications.

Basis for no siqnificant hazards determination:

Pursuant to 10CFR50.92, Vermont Yankee has reviewed the proposed change and concludes that the change does not involve a significant hazards consideration since the proposed change satisfies the criteria in 10CFR50.92(c).

1.

The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a siqnificant increase in the Probability or consequences of an accident Previously evaluated.

The proposed changes do not significantly affect the design or fundamental operation and maintenance of the plant. Accident initiators or the frequency of analyzed accident events are not significantly affected as a result of the proposed changes; therefore, there will be no significant change to the probabilities of accidents previously evaluated.

The proposed changes do not significantly alter assumptions or initial conditions relative to the mitigation 'of an accident previously evaluated. The proposed changes continue to ensure process variables, structures, systems, and components (SSCs) are maintained consistent with the safety analyses and licensing basis.

The revised Technical Specifications continue to require that SSCs are properly maintained to ensure operability and performance of safety functions as assumed in the safety analyses. The design basis events analyzed in the safety analyses will not change significantly as a result of the proposed changes to the TS.

2 NUREG 1433, Revision 3, "Standard Technical Specifications General Electric Plants, BWR/4," dated March 31,2004

Entergy Nuclear Operations, Inc.

Letter Number: BVY 08-007 Vermont Yankee Nuclear Power Station / Page 2 Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not create the Possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involve any physical alteration of the plant (no new or different type of equipment being installed) and do. not involve a change in the design, normal configuration or basic operation of the plant. The proposed changes do not introduce any new accident initiators. In some cases, the proposed changes impose different requirements; however, these new requirements are consistent with the assumptions in the safety analyses.

Where requirements are relocated to other licensee-controlled documents, adequate controls exist to ensure their proper maintenance.

The proposed changes do not involve' significant changes in the fundamental methods governing normal plant operation and do not require unusual or uncommon operator actions. The proposed changes provide assurance that the plant will not be operated in a mode or condition that violates the essential assumptions or initial conditions in the safety analyses and that SSCs remain capable of performing their intended safety functions as assumed in the same analyses. Consequently, the response of the plant and the plant operator to postulated events will not be significantly different.

Therefore, the proposed TS change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

The operation of Vermont Yankee Nuclear Power Station in accordance with the proposed amendment will not involve a significant reduction in a marglin of safety.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. The proposed changes do not significantly affect any of the assumptions, initial conditions or inputs to the safety analyses. Plant design is unaffected by these proposed changes and will continue to provide adequate defense-in-depth and diversity of safety functions as assumed in the safety analyses.

There are no proposed changes to any of the Safety Limits or Limiting Safety System Setting requirements., The proposed changes maintain requirements consistent with safety analyses assumptions and the licensing basis.

Fission product barriers will continue to meet their design capabilities without any significant impact to their, ability to maintain parameters within acceptable limits., The safety functions are maintained within

ýacceptable limits without any significant decrease in capability.

Therefore, the proposed changes do not involve a significant reduction in a margin of saf ety.

Docket No. 50-271 BVY 08-007 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 276 Revision to TS Section 3.3/4.3 for Control Rod System Marked-up Version of the Current Technical Specifications

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS I

With the required shutdown margin not met and the mode switch in the "Refuel" position, immediately suspend Alteration of the Reactor Core except for control rod insertion and fuel assembly removal; immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies; within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , initiate action to restore the integrity of the Secondary Containment System.

2. Reactivity Margin -

Inoperable Control Rods Control rod drives which cannot be moved with control rod drive pressure shall be considered inoperable.

If a partially or fully withdrawn control rod drive cannot be moved with drive or scram pressure, the reactor shall be brought to a shutdown condition within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless investigation demonstrates that the cause of the failure is not due to a failed control rod drive mechanism collet housing.

The control rod directional control valves for inoperable control rods shall be

2.

Reactivity Margin -

Inoperable Control Rods Each partially or fully withdrawn operable control rod shall be UO'tr exercised one notch at least once eachGF4.

This test shall

-Z--

performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in the event power operation is continuing with two or more inoperable control rods or in the event power operation is continuing with one fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out.

The surveillance need not be completed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the number Amendment No. 4-448, 164 8l 81a

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATrON 4.3 SURVEILLANCE REQUIREMENTS OPERATION

,disarmed?

except for control rods which are inoperable because of scram times greater than those specified in Specification 3.3.C.

In no case shall the number of inoperable rods which are not fully inserted be greater than six during power operation.

of inoperable rods has been reduced to less.

than two and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.

trol.Rods The coupling integrity B.

Control Rods

1. Each control rod shall be either coupled to its drive or placed in the inserted position and Sits directional alves disarmed ct'ical1 When removing 17t-6/o; eý control rod drive e

d for inspection and theeactor is in the ref eling mode, this require ent does not apply.

B.

Con 1.

t shall be verified:

[

(a) wlh-'ha ro wiYNhdrawn t/he first

/

Pime subs quent to each ref eling outage r after maint nance, obs ve discerna le re onse of the n clear nstrumentat* n;

however, fo initial r

s when response is not

/

discer ble, subs en r int l

exe ising ofbhese r ds after t Me reactor is Sritical shall be rformedo nto verit the linstrumentation/

rs s;and (b) %When a rod is fully withdrawn, observe' that the rod does not go to the over-travel position.e Prior to r7.

AD~

k co o.

82

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS 1[-311

2.

The Control Rod Drive Housing Support System shall be in place when the Reactor Coolant System is pressurized above atmospheric pressure with fuel in the reactor vessel unless all operable control rods are fully inserted.

3.

While the reactor is below 17% power, the Rod Worth Minimizer (RWM) shall be operating while moving control rods except that:

(a)

If after withdrawal of at least 12 control rods during a startup, the RWM

fails, the startup may continue provided a second licensed operator verifies that the operator at the reactor console is following the control rod program; or (b)

If all rods, except those that cannot be moved with control rod drive

2.

The Control Rod Drive Housing Support System shall be ins ected after rea~ssembly. nd ihe the inspe/tion

ýr L*

on*-z-./

3.

Prior to control rod withdrawal for startup the Rod Worth Minimizer (RWM) shall be verified as operable by performing the following:

r (a)

Verify that the control rod withdrawal sequence for the Rod Worth Minimizer computer is correct.

(b)

The Rod Worth Minimizer diagnostic test shall be performed.

Amendment No.

3-4, 4-4-4, 4-9-6, 8 83

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS pressure, are fully inserted, no more than two rods may be moved.

4.

Control rod patterns and the sequence of withdrawal or insertion shall be established such that the rod drop accident limit of 280 cal/g is not exceeded.

5.

Control rods shall not be withdrawn for startup or refueling unless at least two source range channels have an observed count rate greater than or equal to three counts per second.

(c)

Out-of-sequence

..control rods in each distinct RWM group shall be selected and the annunciator of the selection errors verified.

(d)

An out-of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.

4.

The control rod pattern and sequence of withdrawal or insertion shall be verified to comply with Specification 3.3.B.4.

5.

Prior to control rod withdrawal for startup or during refueling, verification shall be made that at least two source range channels have an observed count rate of at least three counts per second.

I Lf.ý A4re OAD-V e

I 3

--c 4 ;

v 2_4 AC-rJ 0

611r4

6.

Deleted.

I Amendment No.

-9, 6-,

219 84

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS The scram discharge volume drain and vent valves shall be verified open at least once per month.

These valves may be closed intermittently

,(or testing under gdministrative control.

Amendment No. -14, -

-93, -74, 3, 85

VYNPS 4.3 SURVEILLANCE REQUIREMENTS

2.

The maximum scram in r'n o any operable control rod shall not exceed 7.00 seconds;

-Msejc i4 9[91 Amendment No.

-9, 99-T86 86

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION

3.

If Speci ication C,

p-3.3.C.12 cannot be met, fli7 the reac or shall not be made supercritical; if operating, the reactor A",

shall be _u o-wn e

imme~i.ately upon

/

Jno -*det

-mination that Dn avTage s ram tirrl is de' icien

4.

If Specification 3.3.C.2 cannot be met, the deficient control rod shall be considered inoperable, fully min e e

" to the core, A

disarmeý D.

Control Rod Accumulators D.

Control Rod Accumulators At all eactor operating Once a hiftc eck the press es, a rod ac unulator status of the pressure and may b inoperable povided lever alarm for eac tha no other con rol rod in accmulat th nine-rod s: re array ound this ro has a:

1. Inopera le accumulator.

2.

Dire ional control va e electrically eV y 7 5

r;fy d'sarmed while in on-fully insert eciOcA c sr~rc cratv) position.

~c A A

~~5.'

/

Scram insert'on greater s

than maxim permissible insertion ime.

If a contr 1 rod with an inoperab accumulator insertef full-in° and its direc lonal control v Ives are erlectrically di

rmed, it ihall not be co sidered tc/have an iope ble accumulator.

C1 7]

Amendment No. v8 87

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS E.

Reactivity Anomalies E.

Reactivity Anomalies The reactivity equivalent of the difference between the actual critical rod configuration and the expected configuration during power operation shall not exceed 1% Ak/k.

If this limit is exceeded, the reactor will be shut down' until the cause has been determined and corrective actions have been taken if such actions are appropriate.

I Specifi /ations 3.

B hrough 3

.D above are not met, an rderly s tdown shall initiat and the react r shall b in the c d

shu own condi ion withi 24 ours.

During the startup test program and startups following refueling outages, the critical rod configurations will be compared to the expected configurations at selected operating conditions.

These comparisons will be'used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle.

At specific power operating conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately corrected past data.

This comparison will be made at least every equivalent full power month.

cE 19-~

Amendment No.

4, 4-44, 1-4 88

VYNPS 3.12 LIMITING CONDITIONS FOR OPERATION D.

Control Rod and Control Rod Drive Maintenance One control rod may be withdrawn from the core for the purpose of performing control rod and/or control rod drive maintenance provided the following conditions are satisfied:

1. The reactor mode switch shall be locked in the "Refuel" position and the required refueling interlocks shall be operable.

2.

Specification 3.3.A.1 shall be met, or the control rod directional control valves for a minimum of eight control rods surrounding the drive out of service for mainten n

_1A

_be disarme t

alc -c and sufficient margin to criticality demonstrated.

3.

SRMs shall be operable in the core quadrant containing the control rod on which maintenance is being performed and in an adjacent quadrant.

The requirements for an SRM to be considered operable are given in Specification 3.12.B.

4.12 SURVEILLANCE REQUIREMENTS D.

Control Rod and Control Rod Drive Maintenance

1. Prior to performing this maintenance, core shutdown margin shall be determined in accordance with Specification 3.3.A.1 to ensure that the core can be made subcritical at any time during the maintenance with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.

2.

Alternately, if a-minimum of eight control rods surrounding the control rod out of service for maintenance are to be fully inserted and have their directional control valves electrically disarmed, the required shutdown margin shall be met with the strongest control rod remaining in service during the maintenance period fully withdrawn.

Amendment No.

44, -144, 200 232

INSERT 1 [TS 3.3.C.1]

When the reactor is in the STARTUP or RUN MODES;

a. No more than 6 OPERABLE control rods shall be "slow," in accordance with Table 4.3.C-1, and

b. No more than 2 OPERABLE control rods that are "slow" shall occupy diagonally or face adjacent locations.

INSERT2 [TS SR 4.3.C.1]

NOTE:

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.

a. Prior to exceeding 30% RATED THERMAL POWER (RTP) after each reactor shutdown of

- 120 days, verify each control rod scram time is within the limits of Table 4.3.C-1 with reactor steam dome pressure 2_ 800 psig.

b. Every 200 days cumulative operation in RUN MODE, verify, for a representative sample, each tested control rod scram time is within the limits of Table 4.3.C-1 with reactor steam dome pressure > 800 psig.

c. Prior to declaring a control rod OPERABLE after work on a control rod or the CRD System that could affect scram time, verify each affected control rod scram time is within the limits of Table 4.3.C-1 with any reactor steam dome pressure.

d. Prior to exceeding 30% RTP after fuel movement within the affected core cell AND prior to exceeding 30% RTP after work on a control rod or the CRD System that could affect scram time, verify each affected control rod scram time is within the limits of Table 4.3.C-1 with reactor steam dome pressure > 800 psig.

INSERT 3 TABLE 4.3.C-1 Control Rod Scram Times NOTES:

1. OPERABLE control rods with scram times not within the limits of this Table are considered "slow."

2. Follow the Required Actions of LCO 3.3.0.4 for control rods with scram times > 7 seconds to notch position 06. These control rods are inoperable, in accordance with SR 4.3.C.2, and are not considered "slow."

NOTCH POSITION SCRAM TIMES(a)(b)

(seconds)

WHEN REACTOR STEAM DOME PRESSURE 800 psig 46 0.44 36 1.08 26 1.83 06 3.35 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

(b) Scram times as a function of reactor steam dome pressure, when < 800 psig, are within established limits.

INSERT 4 [TS 4.3.C.2]

In accordance with SR's 4.3.C.1.a, b, c & d above, verify each control rod scram time from fully withdrawn to notch position 06 is < 7 seconds.

INSERT 5 [TS 3.3.D]

Each control rod scram accumulator shall be OPERABLE when in the STARTUP or RUN MODES.

NOTE:

Separate action item entry is allowed for each control rod scram accumulator.

1. If a control rod scram accumulator is inoperable with reactor steam dome pressure

_ 800 psig:

NOTE:

TS 3.3.D.1.a only applicable if the associated control rod scram time was within the limits of Table 4.3.C-1 during the last scram time Surveillance.

a.

Declare the associated control rod scram time "slow" within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

- OR -

b.

Declare the associated control rod inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

2. If two or more control rod scram accumulators are inoperable with reactor steam dome pressure _> 800 psig:

a.

Verify / restore the charging water header pressure to > 940 psig within 20 minutes.

- AND -

NOTE:

TS 3.3.D.2.b.1 only applicable if the associated control rod scram time was within the limits of Table 4.3.C-1 during the last scram time Surveillance.

b.1 Declare the associated control rod scram time "slow" within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,

- OR-b.2 Declare the associated control rod inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

3. If one or more control rod scram accumulators are inoperable with reactor steam dome pressure < 800 psig:

a.

Verify/insert all control rods associated with inoperable accumulators immediately upon discovery of charging water header pressure < 940 psig,

- AND -

b.

Declare the associated control rod inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

4.

NOTE:

Not applicable if all inoperable control rodscram accumulators are associated with fully inserted control rods.

If Specifications 3.3.D.2.a or 3.3.D.3.a are not met, ensure all rods are fully inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Docket No. 50-271 BVY 08-007 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 276 Revision to TS Sections 3.3/4.3 for Control Rod System Retyped Technical Specification Pages

Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Letter Number: BVY 08-007 / Page 1 Listing of Affected Technical Specifications Pages:

Replace the Vermont Yankee Nuclear Power Station Technical Specifications pages listed below with the revised pages.

The revised pages contain vertical lines in the margin indicating the areas of change.

Remove Insert 81a

81a, 82 82 83 83 84 84 85 85 86 86 87 87 87a 87b 88 88 232 232

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS With the required shutdown margin not met and the mode switch in the "Refuel" position, immediately suspend Alteration of the Reactor Core except for control rod insertion and fuel assembly removal; immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies; within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , initiate action to restore the integrity of the Secondary Containment System.

2. Reactivity Margin -

Inoperable Control Rods Control rod drives which cannot be moved with control rod drive pressure shall be considered inoperable.

If a partially or fully withdrawn control rod drive cannot be moved with drive or scram pressure, the reactor shall be brought to a shutdown condition within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless investigation demonstrates that the cause of the failure is not due to a failed control rod drive mechanism collet housing.

The control rod directional control valves for inoperable control rods shall be

2. Reactivity Margin -

Inoperable Control Rods Each partially or fully withdrawn operable control rod shall be exercised one notch at least once each month.

This test shall be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in the event power operation is continuing with two or more inoperable control rods or in the event power operation is continuing with one fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out.

The surveillance need not be completed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the number Amendment No. 4-44s, 4-64 a

81a

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION disarmed except for control rods which are inoperable because of scram times greater than those specified in Specification 3.3.C.

In no case shall the number of inoperable rods which are not fully inserted be greater than six during power operation.

B. Control Rods

1. Each control rod shall be either coupled to its drive or placed in the inserted position and its directional valves disarmed.

When removing control rod drives for inspection and the reactor is in the refueling mode, this requirement does not apply.

4.3 SURVEILLANCE REQUIREMENTS of inoperable rods has been reduced to less than two and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.

B.

Control Rods

1. The coupling integrity shall be verified:

(a) When a rod is fully

withdrawn, observe that the rod does not go to the

-over-travel position.

(b) Prior to declaring a control rod-OPERABLE after work on a control rod or CRD system that could affect coupling, each rod shall be fully withdrawn and verified that the rod does not go to the over-travel position.

Amendment No. 2-G 882

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS

2.

The Control Rod Drive Housing Support.System shall be in place when the Reactor Coolant System is pressurized above atmospheric pressure with fuel in the reactor vessel unless all operable control rods are fully inserted.

3.

While the reactor is below 17% power, the Rod Worth Minimizer (RWM) shall be operating while moving control rods except that:

(a)

If after withdrawal of at least 12 control rods during a startup, the RWM

fails, the startup may continue provided a second licensed operator verifies that the operator at the reactor console is following the control rod program; or (b) If all rods, except those that cannot be moved with control rod drive

2.

The Control Rod Drive Housing Support System shall be inspected after reassembly.

3.

Prior to control rod withdrawal for startup the Rod Worth Minimizer (RWM) shall be verified as operable by performing the following:

(a)

Verify that the control rod withdrawal sequence for the Rod Worth Minimizer computer is correct.

(b)

The Rod Worth Minimizer diagnostic test shall be performed.

Amendment No.

3-9,, 4-44, 4-94,

2-2-4 83

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS pressure, are fully inserted, no more than two rods may be moved.

4.

Control rod patterns and the sequence of withdrawal or insertion shall be established such that the rod drop accident limit of 280 cal/g is not exceeded.

5.

Control rods shall not be withdrawn for startup or refueling unless at least two source range channels have an observed count rate greater than or equal to three counts per second.

6.

If the above specifications are not satisfied, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(c)

Out-of-sequence control rods in each distinct RWM group shall be selected and the annunciator of the selection errors verified.

(d)

An out-of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.

4.

The control rod pattern and sequence of

.withdrawal or insertion shall be verified to comply with Specification 3.3.B.4.

5.

Prior to control rod withdrawal for startup or during refueling, verification shall be made that at least two source range channels have an observed count rate of at least three counts per second.

6.

Deleted Amendment No.

4, 4&-,

24-8 84

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS C.

Scram Insertion Times

1. When the reactor is in the STARTUP or RUN MODES;

a.

No more than 6 OPERABLE control rods shall be "slow,"

in accordance with Table 4.3.C-1, AND

b.

No more than 2 OPERABLE control rods that are "slow" shall occupy diagonally or face adjacent locations.

7.

The scram discharge volume'drain and vent valves shall be verified open at least once per month.

These valves may be closed intermittently for testing under administrative control.

C.

Scram Insertion Times NOTE:--------------

During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.

l.a.

Prior to exceeding 30% RATED THERMAL POWER (RTP) after each reactor shutdown of 120 days, verify each control rod scram time is within the limits of Table 4.3.C-1 with reactor steam dome pressure Ž 800 psig.

b.

Every 200 days cumulative operation in RUN MODE, verify, for a representative sample, each tested control rod scram time is within the limits of Table 4.3.C-1 with reactor steam dome pressure

> 800 psig.

c.

Prior to declaring a control rod OPERABLE after work on a control rod or the CRD System that could affect scram time, verify each affected control rod scram time is within the limits of Table 4.3.C-1 with any reactor steam dome pressure.

d.

Prior to exceeding 30% RTP after fuel movement within the affected core cell AND prior to exceeding 30% RTP after work on a control rod or the CRD System that could affect scram time, verify each affected control rod scram time is within the limits of Table 4.3.C-1 with reactor steam dome pressure

> 800 psig.

Amendment No. 4, 2-5, 3-4,

-7G, 7-4, 24-B-85

VYNPS 3.3 LIMITING:CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS TABLE 4.3.C-1 Control Rod Scram Times NOTES--------------

1.

OPERABLE control rods with scram times not within the limits of this Table are considered "slow."

2.

Follow the Required Actions of LCO 3.3.C.4 for control rods with scram times > 7 seconds to notch position 06.

These control rods are inoperable, in accordance with SR 4.3.C.2, and are not considered "slow."

NOTCH SCRAM TIMES (a) (b)

POSITION I (seconds)

WHEN REACTOR STEAM DOME PRESSURE

> 800 psig 46 0.44 36 1.08 26 1.83 06 3.35 (a)

Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

(b) Scram times as a function of reactor steam dome pressure, when < 800 psig, are within established limits.

2.

In accordance with SR's 4.3.C.l.a,b,c & d above, verify each control rod scram time from fully withdrawn to notch position 06 is

< 7 seconds.

2.

The maximum scram insertion time to notch position 06 of any OPERABLE control rod shall not exceed 7.00 seconds.

Amendment No.

9-9,

-4 86

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION

3.

If Specification 3.3.C.l. cannot be met, the reactor shall not be made supercritical; if operating, the reactor shall be placed in the HOT SHUTDOWN condition

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.

If Specification 3.3.C.2 cannot be met, the deficient control rod shall be considered inoperable, fully inserted into the core within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , and disarmed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

D.

Control Rod Accumulators D.

Control Rod Accumulators Each control rod scram Once every 7 days verify accumulator shall be each control rod scram OPERABLE when in the STARTUP accumulator nitrogen or RUN MODES.

pressure is

Ž 940 psig.

NOTE-------------

Separate action item entry is allowed for each control rod scram accumulator.

1. If a control rod scram accumulator is inoperable with reactor steam dome pressure Ž 800 psig:

NOTE-------------

Only applicable if the associated control rod scram time was within the limits of Table 4.3.C-1 during the last scram time Surveillance.

a.

Declare the associated control rod scram time "slow" within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

-OR-

b. Declare the associated control rod inoperable within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Amendment No. 8 87

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION

2.

If two or more control rod scram accumulators are inoperable with reactor steam dome pressure Ž 800 psig:

a.

Verify/restore the charging water header pressure to

> 940 psig within 20 minutes.

-AND-NOTE-------------

Only applicable if the associated control rod scram time was within the limits of Table 4.3.C-1 during the last scram time Surveillance.

b.l Declare the associated control rod scram time "slow" within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,

-OR-b.2 Declare the associated control rod inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

3.

If one or more control rod scram acculumators are inoperable with reactor steam dome pressure < 800 psig:

a.

Verify all control rods associated with inoperable accumulators are fully inserted immediately upon discovery of charging water header pressure

< 940 psig.

-AND-

b.

Declare the associated control rod inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Amendment No.

87a

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION

4.

If Specifications 3.3.D.2.a or 3.3.D.3.a are not met, ensure all rods are fully inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

NOTE-------------

The above specification is not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods.

Amendment No.

87b

VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS E.,

Reactivity Anomalies The reactivity equivalent of the difference between the actual critical rod configuration and the expected configuration during power operation shall not exceed 1% Ak/k.

If this limit is exceeded, the reactor will be shut down until the cause has been determined and corrective actions have been taken if such actions are appropriate.

E.

Reactivity Anomalies During the startup test program and startups following refueling outages, the critical rod configurations will be compared to the expected configurations at selected operating conditions.

These comparisons will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle.

At specific power operating conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately corrected past data.

This comparisonwill be made at least every equivalent full power month.

Amendment No.

-44, 4-4-&, 8 88

VYNPS 3.12 LIMITING CONDITIONS FOR OPERATION D. Control Rod and Control Rod Drive Maintenance One control rod may be withdrawn from the core for the purpose of performing control rod and/or control, rod drive maintenance provided the following conditions are satisfied:

1. The reactor mode switch shall be locked in the "Refuel" position and the required refueling interlocks shall be operable.

2. Specification 3.3.A.1 shall be met, or the control rod directional control valves for a minimum of eight control rods surrounding the drive out of service for maintenance shall be disarmed and sufficient margin to criticality demonstrated.

3. SRMs shall be operable in the core quadrant containing the control

,rod on which maintenance is being performed and in an adjacent quadrant.

The requirements for an SRM to be considered operable are given in Specification 3.12.B.

4.12 SURVEILLANCE REQUIREMENTS D. Control Rod and Control Rod Drive Maintenance

1. Prior to performing this maintenance, core shutdown margin shall be determined in accordance with Specification 3.3.A.1 to ensure that the core can be made subcritical at any time during the maintenance with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.

2. Alternately, if a minimum of eight control rods surrounding the control rod out of service for maintenance are to be fully inserted and have their directional control valves electrically disarmed, the required shutdown margin shall be met with the strongest control rod remaining in service during the maintenance period fully withdrawn.

Amendment No. 4-s-,

4-4-9, 2-9423 232

Docket No. 50-271 BVY 08-007 Vermont Yankee Nuclear Power Station Proposed Change No. 276 Revision to TS Section 3.3/4.3 for Control Rod System Proposed Revision to TS Bases Pages (for information only)

VYNPS BASES:

3.3 & 4.3 CONTROL ROD SYSTEM A.

Reactivity Limitations

1. Reactivity Margin -

Core Loading The specified shutdown margin (SDM) limit accounts for the uncertainty in the demonstration of SDM by testing.

Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement.

This is due to the reduced uncertainty in the SDM test when the highest worth control rod is determined by measurement (e.g.,

SDM may be demonstrated by an in-sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or by local criticals, where the highest worth rod is determined by testing).

Following a refueling, adequate SDM must be demonstrated to ensure that the reactor can be made subcritical at any point during the cycle.

Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle.

Therefore, to obtain the SDM, the initial measured value must exceed LCO 3.3.A.1 by an adder, "R",

which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity.

If the value of "R" is negative (that is, BOC is the most reactive point in the cycle),

no correction to the BOC measured value is required.

The value of R shall include the potential shutdown margin loss assuming full B4C settling in all inverted poison tubes present in the core.

The frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.

When SDM is demonstrated by calculations not associated with a test (e.g., to confirm SDM during the fuel loading sequence),

additional margin must be included to account for uncertainties in the calculation.

During refueling, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals.

An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling.

This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern.

For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence.

These bounding analyses include additional margins to account for the associated uncertainties in the calculation.

Amendment No.

L-4, NVY 87 131, l-4-88 89

VYNPS BASES:

3.3 & 4.3 (Cont'd)

2.

Reactivity Margin -

Inoperable Control Rods Specification 3.3.A.2 *requires that a rodhbe taken out of service if it cannot be moved with drive pressure.

Ifa rod is

disarmed, its position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A.1'.

This assures that the core can be shutdown at all times with the remaining control rods, assuming the highest worth, operable control rod does not insert.

An-allowable pattern for control rods valved out of service will be available to the reactor operator.

The number of rods permitted to be inoperable could be many more than the six allowed by the Specification, particularly late in the operation cycle; however, the occurrence of more than six could be indicative of a generic control rod drive problem and the reactor will be shutdown.

Also if damage within the control rod drive mechanism and in particular, cracks in drive internal housing, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled out.

Circumferential cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs.

This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods.

Limiting the period of operation with a potentially severed collet housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collet housings.

The monthly control rod exercise test serves as a periodic check against deterioration of the Control Rod System and also verifies the ability of the control rod drive to scram.

The frequency of exercising the control rods under the conditions of two or more control rods valved out of service provides even further assurance of the reliability of the remaining control rods.

B.

Control Rods

1. Control rod dropout accidents as discussed in the UFSAR can lead to significant core damage.

If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated.

Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary.

The Surveillance requires verifying a control rod does not go to the withdrawn over-travel position.

The over-travel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the over-travel position.

The verification is required to be performed any time a control rod is withdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling.

This includes control rods inserted one notch and then returned to the "full out" position during the performance of SR 4.3.A.2.

This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events.

Amendment No.

4-44, 1-4-9 89a

VYNPS BASES:

3.3 & 4.3 (Cont'd)

2.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure.

The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of the primary coolant system.

The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation is given in Subsection 3.5.4.

This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

1

3.

In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is followed.

Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed.

The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional check that the reactor operator is following prescribed sequence.

Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails subsequently is acceptable if a second licensed operator verifies the withdrawal sequence.

Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod.

Withdrawal of rods for testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully inserted.

Above 17% power, the RWM is not needed since even with a single error an operator cannot withdraw a rod with sufficient worth, which if dropped, would result in any-thing but minor consequences.

4.

Refer to the "General Electric Standard Application for Reactor Fuel (GESTAR II),"

NEDE-24011-P-A, (the latest NRC-approved version will be listed in the COLR).

5.

The Source Range Monitor (SRM) system provides a scram function in noncoincident configuration.

It does provide the operator with a visual indication of neutron level.

The consequences of reactivity accidents are a function of the initial neutron flux.

The requirement of at least three counts per second assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions.

One operable SRM channel is adequate to monitor the approach to criticality, therefore, two operable SRM's are specified for added conservatism.

6.

The action statement for TS 3.3.B.6 requires that the plant be placed in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if the required actions of TS 3.3.B.1 through 3.3.B.5 are not satisfied.

This ensures that all insertable control rods are inserted and places. the reactor in a

condition that does not require the active function (i.e.,

scram) of the control rods.

The allowed completion time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based upon operating experience to reach HOT SHUTDOWN from full power in an orderly manner and without challenging plant systems.

Amendment No.

L-5, 4-,

4-4.4, BVY 99 ill, BVY04-1 4G, _9 90

VYNPS BASES:

3.3 & 4.3 (Cont'd)

7.

Periodic verification that the Scram Discharge Volume (SDV) drain and vent valves are maintained in the open position provides assurance that the SDV will be available to accept the water displaced from the control rod drives in the event of a scram.

C.

Scram Insertion Times BACKGROUND The scram function of the Control Rod Drive (CRD)

System controls reactivity changes during abnormal operational transients to ensure that specified acceptable fuel design limits are not exceeded.

The control rods are scrammed by positive means using hydraulic pressure exerted on the CRD piston.

When a scram signal is initiated, control air is vented from the scram valves, allowing them to open by spring action.

Opening the exhaust valve reduces the pressure above the main drive piston to atmospheric

pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston.

Since the notches in the index tube are tapered on the lower edge, the collet fingers are forced open by cam action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston.

As the drive moves upward and the accumulator pressure reduces below the reactor pressure, a ball check valve opens, letting the reactor pressure complete the scram action.

If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod in the required time without assistance from reactor pressure.

APPLICABLE SAFETY ANALYSES The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate.

The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., MCPR).

Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity.

Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met.

The scram function of the CRD System protects the MCPR Safety Limit (SL)

(reference TS 1.1.A, "Bundle Safety Limit (Reactor Pressure >800 psia and Core Flow >10% of Rated)," and TS 3.1l.C, "Minimum Critical Power Ratio (MCPR)")

and the 1% cladding plastic strain fuel design limit (reference specification 3.1l.A, "Average Planar Linear Heat Generation Rate (APLHGR)"),

which ensure that no fuel damage will occur if these limits are not exceeded.

Above 800 psig, the scram function is designed to insert negative, reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient.

Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Reference 1) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (Reference TS 3.3.B.3 and 3.3.B.4, regarding the Rod Worth Minimizer and control rod patterns).

For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of 10 CFR 50.36(c) (2)

(ii).

Amendment No.

2-5, q4, 4-44, BVY 99 !1!,

Blj04G1 40 91

VYN PS BASES:

3.3 & 4.3 (Cont'd)

LCO The scram times specified in Table 4.3.C-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Reference 2).

To account for single failures and "slow' scramming control rods, the scram times specified in Table 4.3.C-1 are faster than those assumed in the design basis analysis.

The scram times have a margin that allows up to approximately 7% of the control rods (e.g., 89 x 7%

6) to have scram times exceeding the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as limited by TS 3.3.A.

"Reactivity Limitations") and an additional control rod failing to scram per the single failure criterion.

The scram times are specified as a function of reactor steam dome pressure-to account for the pressure dependence of the scram times.

The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication.

The reed switch closes ("pickup")

when the index tube passes a specific location and then opens

("dropout") as the index tube travels upward.

Verification of the specified scram times in Table 4.3.C-1 is accomplished through measurement of the "dropout" times.

To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.

Table 4.3.C-1 is modified by two Notes which state that control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 4. *3.C.2. Slow scramming control rods may be conservatively declared inoperable and not accounted for as "slow" control rods.

APPLICABILITY In STARTUP and RUN MODES, a scram is assumed to function during transients and accidents analyzed for these plant conditions.

These events are assumed to occur during startup and power operation; therefor 'e, the scram function of the control rods is required during these MODES.

In SHUTDOWN, the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied.

This provides adequate requirements for control rod scram capability during these conditions.

In REFUELING, only one control rod is able to be withdrawn.

Additional restrictions and requirements *when in REFUELING can be found in TS 3.12 "Refueling and Spent Fuel Handling."

REQUIRED ACTIONS TS 3.3.C.3 When the requirements of TS 3.3.C.1 are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analyses.

Therefore, the plant must be brought to a MODE in which the LCD does not apply.

To achieve this status, the plant must be brought to at least the HOT SHUTDOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed completion time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach the SHUTDOWN MODE from full power conditions in an orderly manner and without challenging plant systems.

Amendment No.

a 91a

VYNPS BASES:

3.3 & 4.3 (Cont'd)

TS 3.3.C.4 Specification 3.3.C.2 requires that no operable control rod have a scram time greater than 7 seconds.

TS 3.3.C.4 requires that for control rods that do not satisfy the 7 second requirement, that they be considered inoperable.

In addition, the subject control rod must be fully inserted into the core within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and (electrically or hydraulically) disarmed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.

The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations.

The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves.

The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids.

The allowed completion times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS (SR)

The four surveillances of SR 4.3.C.1 are modified by a Note stating that during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator.

With the CRD pump isolated, (i.e.., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times.

During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.

SR 4.3.C.I.a The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time.

Measurement of the scram times with reactor steam dome pressure > 800 psig demonstrates acceptable scram times for the transients analyzed.

Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy.

Therefore, demonstration of adequate scram times at reactor steam dome pressure

> 800 psig ensures that the measured scram times will be within the specified limits at higher pressures.

Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed.

To ensure that scram time testing is performed within a reasonable time following a shutdown

> 120 days or longer, control rods are required to be tested before exceeding '30% RTP following the shutdown.

This frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.

Amendment No.

91b

VYNPS BASES:

3.3 & 4.3 (Cont'd)

SR 4.3.C.i.b Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle.

A representative sample contains at least 10% of the control rods.

The sample remains representative if no more than 7.5% of the control rods in the sample tested are determined to be "slow."

With more than 7.5%

of the sample declared to be "slow" per the criteria in Table 4.3.C-1, additional control rods are tested until this 7.5% criterion (e.g.,

7.5% of the entire sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all surveillances) exceeds the LCO limit.

For planned testing, the control rods selected for the sample should be different for each test.

Data from inadvertent scrams should be used whenever -possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample.

The 200 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle.

This Frequency is also reasonable based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with SR 4.3.A.2 "Notch Testing" and SR 4.3.D, "Control Rod Accumulators."

SR 4.3.C.l.c When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure.

The scram testing must be performed once before declaring the control rod OPERABLE.

The required scram time testing must demonstrate the affected control rod is still within acceptable limits.

The limits for reactor pressures < 800 psig are established based on a high probability of meeting the acceptance criteria at reactor pressures >- 800 psig.

Limits for -> 800 psig are found in Table 4.3.C-1.

If testing demonstrates the affected control rod does not meet these limits, but is within the 7 second limit of Table 4.3.C-1, Note 2, the control rod can be declared OPERABLE and "slow."

Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram.

The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

Amendment No.

91c

VYNPS BASES:

3.3 & 4.3 (Cont'd)

SR 4.3.C.l.d When work that could affect the scram insertion time is performed on a control rod or CRD System, or when fuel movement within the reactor pressure vessel occurs, testing must be done to demonstrate each affected control rod is still within the limits of Table 4.3.C-1 with the reactor steam dome pressure >- 800 psig.

Where work has been performed at high reactor pressure, the requirements of SR 4.3.C.l.c and SR 4.3.C.l.d can be satisfied with one test.

For a control rod affected by work performed while shut down,

however, a zero pressure and high pressure test may be required.

This testing ensures that, prior to withdrawing the control rod for continued operation; the control rod scram performance is acceptable for operating reactor pressure conditions.

Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria.

When fuel movement within the reactor pressure vessel occurs, only those control rods associated with the core cells affected by the fuel movement are required to be scram time tested.

During a routine refueling outage, it is expected that all control rods will be affected.

The Frequency of once prior to exceeding 30% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.

SR 4.3.C.2 Verifying that the scram time for each control rod to notch position 06 is

-< 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.

This SR is performed in conjunction with the control rod scram time testing of SR 4.3.C.l.a, SR 4.3.C.i.b, SR 4.3.C.l.c, and SR 4.3.C.l.d.

The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.

REFERENCES

1. NEDE-24011-P-A-9, "General Electric Standard Application for Reactor Fuel," Section 3.2.4.1, September 1988.

2.

Letter from R.F. Janecek (BWROG) to R.W.

Starostecki (NRC),

"BWR Owners Group Revised Reactivity Control System Technical Specifications,"

BWROG-8754, dated September 17, 1987.

D.

Control Rod Accumulators BACKGROUND The control rod scram accumulators are part of the Control Rod Drive (CRD)

System and are provided to ensure that the control rods scram under varying reactor conditions.

The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure.

The accumulator is a hydraulic cylinder with a free floating piston.

The piston separates the water used to scram the control rods from the nitrogen, which provides the required energy.

The scram accumulators are necessary to scram the control rods within the required insertion times of LCO 3.3.C, "Scram Insertion Times."

Amendment No.

91d

VYNPS BASES:

3.3 & 4.3 (Cont'd)

APPLICABLE SAFETY ANALYSES The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate.

OPERABILITY of each individual control rod scram accumulator, along with LCO 3.3.A.2, "Reactivity Margin -

Inoperable Control Rods,"

LCO 3.3.B "Control Rods," and LCO 3.3.C "Scram Insertion Times", ensures that the scram reactivity assumed in the DBA and transient analyses can be met.

The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated control rod.

The scram function of the CRD System, and therefore the OPERABILITY of the accumulators, protects the MCPR Safety Limit (reference TS 1.1.A, "Bundle Safety Limit (Reactor Pressure >800 psia and Core Flow >10% of Rated)," and TS 3.1l.C, "Minimum Critical Power Ratio (MCPR)")

and 1%

cladding plastic strain fuel design limit (reference specification 3.1l.A, "Average Planar Linear Heat Generation Rate (APLHGR),")

and TS 3.1l.B, "Linear Heat Generation Rate (LHGR)"),

which ensure that no fuel damage will occur if these limits are not exceeded.

In addition, the scram function at low reactor vessel pressure (i.e.,

startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (Reference TS 3.3.B.3 and 3.3.B.4, regarding the Rod Worth Minimizer and control rod patterns).

Control rod scram accumulators satisfy Criterion 3 of 10 CFR 50.36(c) (2) (ii).

LCO The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor pressures.

The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure.

APPLICABILITY In STARTUP and RUN MODES, the scram function is required for mitigation of DBAs and transients, and therefore the scram accumulators must be OPERABLE to support the scram function.

In

SHUTDOWN, control rods are not allowed to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied.

This provides adequate requirements for control rod scram accumulator OPERABILITY during these conditions.

In REFUELING, only one control rod is able to be withdrawn.

Additional restrictions and requirements when in REFUELING can be found in TS 3.12 "Refueling and Spent Fuel Handling."

REQUIRED ACTIONS The required actions of TS 3.3.D is modified by a Note indicating that a separate condition entry is allowed for each control rod scram accumulator.

This is acceptable since the required actions for each condition provide appropriate compensatory actions for each inoperable accumulator.

Complying with the Required Actions may allow for continued operation.

Amendment No.

91e

VYN PS BASES:

3.3 & 4.3 (Cont'd)

TS 3.3.D.l.a and 1.b With one control rod scram accumulator inoperable and the reactor steam dome pressure Ž! 800 psig, the control rod m~ay be declared "slow," since the control rod will still scram at the reactor operating pressure but may not satisfy the required scram times in Table 4.3.C-1.

Required action l.a is modified by a Note indicating that declaring the control rod "slow" only applies if the associated control scram time was *within the limits of Table 4.3.C-1 during the last scram time test.

Otherwise, the control rod would already be considered "slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times.

In this event, the associated control rod is declared inoperable (required action 1.b) and LCO 3.3.C.4 is entered.

This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function.

The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on the large number of control rods available to provide the scram function and the ability of the affected control rod to scram only with reactor pressure at high reactor pressures.

TS 3.3.D.2.a, 2.b.1 and 2.b.2 With two or more control rod scram accumulators inoperable and reactor steam dome pressure Ž! 800 psig, adequate pressure must be supplied to the charging water header.

With inadequate charging water header pressure, all of the accumulators could become inoperable, resulting in a potentially severe degradation of the scram performance.

Therefore, within 20 minutes from discovery of charging water header pressure <

940 psig concurrent with condition 2, adequate charging water header pressure must be restored.

The allowed completion time of 20 minutes is reasonable, to place a CR0 pump into service to restore the charging header pressure, if required.

This completion time is based on the ability of the reactor pressure alone to fully insert all control rods.

The control rod may be declared "slow," since the control rod will still scram using only reactor pressure, but may not satisfy the times in Table 4.3.C-1. Required action 2.b.1 is modified by a Note indicating that declaring the control rod "slow" only applies if the associated control scram time is within the limits of Table 4.3.C-1 during the last scram time test.

Otherwise, the control rod would already be considered "slow" and the further degradation of scram performance with an inoperable accumulator could resul~t in excessive scram times.

In this event, the associated control rod is declared inoperable (required action 2.b.2) and LCO 3.3.C.4 entered.

This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function.

The allowed completion time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable.

Amendment No.9f 91f

VYNPS BASES:

3.3 & 4.3 (Cont'd)

TS 3.3.D.3.a and 3.b With one or more control rod scram accumulators inoperable and the reactor steam dome pressure < 800 psig, the pressure supplied to the charging water header must be adequate to ensure that accumulators remain charged.

With the reactor steam dome pressure < 800 psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely degraded during a depressurization event or at low reactor pressures.

Therefore, immediately upon discovery of charging water header pressure

< 940 psig, concurrent with condition 3, all control rods associated with inoperable accumulators must be verified to be fully inserted.

Withdrawn control rods with inoperable accumulators may fail to scram under these low pressure conditions.

The associated control rods must also be declared inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The allowed completion time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable for required action 3.b, considering the low probability of a DBA or transient occurring during the time that the accumulator is inoperable.

TS 3.3.D.4 The reactor must be shutdown within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if either required action and associated completion time associated with loss of the CRD charging pump (required actions 2.a and 3.a) cannot be met.

Shutting down the reactor ensures that all insertable control rods are inserted and that the reactor would then be in a condition that does not require the active function (i.e.,

scram) of the control rods.

This required action is modified by a Note stating that the action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.

SURVEILLANCE REQUIREMENTS SR 4.3.D SR 4.3.D requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator gas pressure exists to provide sufficient scram force.

The primary indicator of accumulator OPERABILITY is the accumulator gas pressure.

A minimum accumulator gas pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable.

The minimum accumulator gas pressure of 940 psig is well below the expected pressure of 1100 psig.

Declaring the accumulator inoperable when the minimum gas pressure is not maintained ensures that significant degradation in scram times does not occur.

The 7 day frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

Amendment No.

91g

(~,

VYNPS BASES:

3.3 & 4.3 (Cont'd)

E.

Reactivity Anomalies During each fuel cycle, excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.

The magnitude of this excess reactivity may be inferred from the critical rod configuration.

As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state.

Power operation base conditions provide the most sensitive and directly interpretable data relative to core reactivity.

Furthermore, using power operating base conditions permits frequent reactivity comparisons. Reactivity anomaly is used as a measure of the predicted versus measured core reactivity during power operation.

If the measured and predicted rod density for identical core conditions at BOC do not reasonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used to predict rod density may not be accurate.

If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured value. Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1%

Ak/k.

Deviations in core reactivity greater than 1% Ak/k are not expected and require thorough evaluation.

One percent reactivity limit is cbnsidered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the Reactor System.

Amendment No.

91h

v t e Letter Sequence Request
TAC:MD8035, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Questions 12 March 2008 25 March 2008 Answers 24 April 2008 Results Approval, Approval Other: BVY 09-022, Update to Spent Fuel Management Plan
MONTHYEAR2008-03-25 [Table View]

BVY 08-007, Technical Specification Proposed Change No. 276 Revision to TS Sections 3.3/4.3 for Control Rod System

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BVY 08-007, Technical Specification Proposed Change No. 276 Revision to TS Sections 3.3/4.3 for Control Rod System

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Dear Sir or Madam:

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